1. Field of the Invention
This invention pertains to the field of nuclear reactors, particularly of the pressurized water type and is concerned with the fluid systems important to safety which mitigate the consequences of any postulated event as required by the Federal Code of Regulations, Title 10. Specifically, these systems provide emergency cooling of the reactor core, reactor building heat removal and pressure reduction, and reactor building fission product control. These fluid systems in conjunction with specific plant layout considerations together with other equipment typical in current reactor plant designs comprise the engineered safeguards features of the nuclear reactor. This invention does not specifically consider those portions of the engineered safety features which are related to the control rods or to instruments and associated electrical components which monitor reactor operation or provide signals for actuation of the engineered safety features although their use is assumed and utilized in the application of these fluid systems.
These fluid systems important to safety in a nuclear reactor stand by to perform the basic functions of heat removal and water inventory control for "events" which result due to failure of normally operating components or systems or result in a situation where necessary compensation lies beyond the capacity or capability of the systems designed for the normal reactor operation. Typical of such events are cracks or breaks at any location of the primary cooling water circuit which may range from a hairline crack in a small pipe to a "guillotine" break in a main pipe, causing water losses which may range from slow leaks to cascading outflows from the primary circuit onto the containment floor. Accordingly, the intensity and rapidity of the development of heat to be removed or coolant to be resupplied may vary greatly dependent on the rate of coolant loss.
These systems important to safety are designed to automatically initiate and perform measures commensurate with the seriousness of the "event", in accordance with established Federal regulations and nuclear industry standards.
2. Description of the Prior Art
Conventional safety systems that are employed in pressurized water reactors are typically dependent on a large number of "active" components such as pumps and fans which have the capability to be powered by emergency diesel engines and associated electrical power systems should off-site power not be available. Thus, for example, assuming a large pipe break in the coolant system, termed as a "loss of coolant accident", water is introduced into the primary circuit and thus the reactor vessel by safety systems comprised of pumps. Further, the containment heat-up is counteracted by pump-operated spraying devices and also, heat from the containment is removed by motor-driven fan coolers. The reactor decay heat and heat from the containment is transferred into an emergency cooling water system which also utilizes pumps. Additional active "safety grade" systems are also required for other safety functions and include an emergency steam generator feed water system as well as ventilating and air conditioning arrangements for cooling these active components. These safety systems are all required to be redundant in order that the failure of any single component will not result in the loss of a safety function. Also, to assure the operation of active components in safety systems, nuclear power plants have required physical separation for fire and flood protection between redundant trains of equipment. This approach to nuclear plant safety results in plant designs which are highly complex and expensive, requiring significant testing, construction and operational efforts associated with direction, maintenance and quality assurance.